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EURATOM/CCFE Fusion Association

Annual Report 2014/15

EURATOM/CCFE Fusion Association Annual Report 2014/15

Directors foreword 1. Introduction2. European and UK fusion research3. Overview of activities in the CCFE

programme4. Tokamak physics 5. MAST Upgrade6. JET operations7. Materials research8. Technology studies9. RACE10. ITER systems11. Outreach, training and industry

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Contents

1EURATOM/CCFE Fusion Association Annual Report 2014/15

Directors foreword

While we continue to play a leading role in the development of fusion energy, we are increasingly finding industrial applications of our novel technology. Culhams work in maintaining and upgrading JET on behalf of our European fusion partners ensures it remains the leading fusion experiment in the world and the most effective test bed for its successor ITER. We are refurbishing the machine and preparing our engineering staff to use the fusion fuel (specifically tritium in addition to deuterium) in JET experiments at the end of this decade. This will allow us to push towards new record performance on JET and just as importantly train a new generation of scientists in running fusion devices which make significant fusion power.

Our own UK domestic programme, centred on the MAST fusion experiment, is at an important stage. The major upgrade to MAST is very well advanced, indeed we are now starting the rebuild of the machine and operation will commence in 2017. This will enable our scientists to study near-fusion grade plasmas at a fraction of the size of JET. MAST Upgrade will ensure that Culham continues to operate a world-leading tokamak experiment through the 2020s.

The growth of our technology activities has resulted in two new centres of excellence. The construction of the first centre, the new Materials Research Facility (MRF), is well underway and will be ready to start operations later this year. It is a key part of the National Nuclear Users Facility, which is being developed at three locations (CCFE, the National Nuclear Laboratory and the Dalton Cumbria Facility of the University of Manchester). The MRF will provide essential facilities for universities, industry and other nuclear research laboratories to develop advanced materials for nuclear power.

The second new centre, RACE (Remote Applications in Challenging Environments), is also under construction. Funded by the UK Governments City Deal initiative in partnership with leading laboratories and industry, RACE enables our considerable expertise in remote handling techniques (honed on JET over the last twenty years) to be available for industry helping them exploit a multi-billion pound world-wide market in robotics and autonomous systems. Indeed, RACE has already helped UK industry obtain over 130m of contracts this year. MRF and RACE are just two examples of a range of new technology initiatives designed to build on our prominent position in fusion research today and make significant contributions to bringing the design of the demonstration fusion power station (DEMO) to full commercialisation.

Without doubt the most enjoyable part of my experience at CCFE is working with some of the most creative and skilled scientists and engineers on the planet. The grand challenge of harnessing fusion power for the world demands invention and innovation like never before, and that comes from exceptional people with extraordinary ideas. People like our new graduate engineers and physicists; like the ever larger number of PhD students we host at Culham and, of course, our award-winning apprentices. These are the people that will be solving the material challenges for fusion and fission power plants, preparing JET for record-breaking performance and running the first plasmas on ITER.

Training these people is crucial and plans are well advanced to build an advanced skills training facility at Culham to train our own apprentices as well as those for other high tech companies in Oxfordshire. Nurturing and empowering the next generation of scientists and engineers is essential for our mission and indeed for the UK.

Professor Steve CowleyHead of EURATOM/CCFE Fusion Association

Culham Centre for Fusion Energy is buzzing with activity; we are assembling an innovative fusion experiment, MAST Upgrade; we are preparing the Joint European Torus (JET) for fusion power operation and we are constructing two new national centres of excellence in remote handling and nuclear materials.

2 EURATOM/CCFE Fusion Association Annual Report 2014/15

1. Introduction and fusion basics

This report covers fusion research at Culham Centre for Fusion Energy (CCFE)1 for the year 2014/15. The objective of fusion research is to develop power stations that harness the process that powers the sun for clean electricity generation here on earth. Fusion power stations will emit no greenhouse gases, their fuel will be abundant and widespread, and their waste products will be much more manageable than that from todays fission nuclear power stations.

In fusion reactions, nuclei of light atoms join together to create nuclei of heavier atoms. The products of the reaction are lighter than the sum of the incoming nuclei, and the mass (m) thus destroyed is converted to energy via Einsteins famous equation, E=mc2 (with c being the value of the speed of light). The reaction products thus have much more energy than the original atoms. The deuterium-tritium (D-T) fusion reaction is the easiest to initiate and is illustrated in Figure 1.1; as well as a helium nucleus it produces a neutron. To achieve copious fusion reactions on earth requires temperatures exceeding 100 million oC (which are routinely reached in JET). At these temperatures the hot gas (plasma) must be kept away from material surfaces. In JET, MAST and ITER this is achieved using magnetic fields. These machines are all types of tokamaks; other designs of magnetic bottle are also studied, for example stellarators. There is an alternative inertial fusion approach in which the fuel is compressed very rapidly with intense laser beams and the plasma lasts only a tiny fraction of a second.

Figure 1.1: The deuterium - tritium fusion reaction produces a helium nucleus (alpha particle) and a neutron. Deuterium and tritium are heavy and super-heavy hydrogen.

In a fusion power station the fast helium from the D-T reaction would remain in the magnetic bottle and so keep the plasma hot, while the neutron being neutral would escape and be captured in a blanket around the plasma. This blanket would (a) get hot - this heat would be used to generate electricity, and (b) contain lithium, which would react with the neutron to create the tritium needed for the fuel (the deuterium is easily extracted from water). As the energy released is about ten million times that from a chemical reaction, the amount of fuel required is correspondingly less; the deuterium in half a bath of water plus the lithium in one laptop battery would provide the fuel for around 200,000 kilowatt-hours of electricity, equal to the UKs per capita electricity consumption for 30 years. 70 tonnes of fuel is required to produce this in a coal-fired power station.

1CCFE is the research arm of the United Kingdom Atomic Energy Authority

Deuterium

Helium

D + T 4He + n + Energy

Tritium

Neutron

+ energy (17.6 MeV)

The Fusion processreaction for first fusion power plants

2. European and UK fusion research

CCFE has a domestic fusion programme funded by a grant from the UK Engineering and Physical Sciences Research Council and also participates strongly in a coordinated European programme. In Europe, fusion research is co-ordinated by EURATOM and comprises:

a) Collective activities managed by EUROfusion2 (https://www.euro-fusion.org/). The EUROfusion programme is focussed on delivering the fusion roadmap (https://www.euro-fusion.org/eurofusion/the-road-to-fusion-electricity/), which includes programmes in support of designing a prototype reactor (DEMO) and programmes in support of ITER including experiments on JET (Joint European Torus, located at Culham). JET is presently the worlds leading magnetic fusion research facility. CCFE has a contract from EURATOM to operate JET, with a programme of experiments by Task Forces of scientists from all EUROfusion members (including CCFE).

b) International collaborations, dominated by ITER (http://www.iter.org), construction of which is well underway in Cadarache in southern France. The partners are EURATOM, China, Japan, India, the Russian Federation, South Korea and the US. ITER is mainly being procured by Domestic Agencies in the partners; Europes is in Barcelona and known as Fusion for Energy (F4E, http://fusionforenergy.europa.eu). The aim of ITER is to produce energy from fusion on a power-plant-relevant scale (500 MW) and test key technologies for power stations like superconducting coils, remote handling, high heat flux components, and the blankets that will capture heat released by fusion reactions. F4E also manages European contributions to the Broader Approach collaboration with Japan concerning fusion projects needed in addition to ITER (including IFMIF the planned International Fusion Materials Irradiation Facility to qualify materials for DEMO, a new JET-scale superconducting tokamak JT-60SA being built in Japan in support of ITER and the design of DEMO, and a supercomputer based in Japan that started operating in December 2011).

Figure 2.1: Progress on the ITER tokamak complex (in February 2015) ( ITER Organization, http://www.iter.org).

2 EUROfusion is a consortium of all the main European fusion institutes who hold an H2020 grant on fusion research


2. European and UK fusion research 2. European and UK fusion research continued...

It is intended that ITER will build up to high fusion power D-T experiments in the latter part of the 2020s, and that a beam driven neutron source (14MeV) for testing material properties is built and operated in parallel, then a demonstration power station could provide electricity to the grid before 2050. Other facilities will be needed to strengthen theprogramme and reduce risk. A major facility under consideration in some countries is a Component Test Facility (CTF); this might be based on the compact spherical tokamak (ST) concept pioneered at Culham through experiments on MAST and its predecessor START. MAST is presently undergoing a large upgrade that commenced in October 2013 and will end inlate 2016.

UK universities make substantial contributions to magnetic confinement studies, in materials science, plasma physics and technology/engineering. Over twenty universities make contributions to research in this area, and many of these have PhD students working on fusion projects, either at Culham or at the university. Much of the programme is undertaken in collaboration with other EUROfusion members and with fusion institutes in the rest of the world.

Figure 2.2: MAST and various internal assemblies during the major upgrade to its capabilities.


3. Overview of activities in the CCFE programme

CCFE has a broad-ranging programme of activities encompassing technology developments for DEMO and power plants, the development of materials suitable for a fusion environment, engineering activities, tokamak plasma physics, the training of students/apprentices, and public and industry information outreach activities. CCFE contributes to all the key areas of study in tokamak plasma physics research, as part of EUROfusion programmes, through experiments on MAST, participation in the JET programme, and plasma theory and modelling. CCFE also operates JET on behalf of EURATOM for use by scientists conducting experiments under EUROfusion. Sections 4 and 6 discuss the tokamak science programme and JET operations, respectively. Currently the MAST experiment is undergoing a major upgrade to its capabilities including the innovative Super-X divertor concept this is covered in section 5.

CCFE also has strong programmes on understanding and improving the materials and technology needed for DEMO and fusion power stations, which are respectively discussed in sections 7 and 8. A major development in the materials area is the new Materials Research Facility (MRF), that is scheduled to complete construction in Autumn 2015. The MRF, which is part of the National Nuclear User Facility, is also discussed in section 7.

A second large facility under construction on the Culham site at present is the RACE (Remote Applications in Challenging Environments) facility. Under the RACE umbrella all remote handling activities, conducted by UKAEA, will be channelled. RACE is discussed in section 9.

Technology development for ITER concentrates on remote handling applications, developing specialist heating systems and instrumentation for plasma measurements (diagnostics). These activities are discussed in section 10.

CCFE recognises the importance of informing its stakeholders of the work it carries out. Accordingly a variety of public events are held on and off the Culham site, including for schoolchildren, VIP visits are facilitated and a large range of media, website and social media output is maintained. CCFE also informs UK companies of opportunities for business in the fusion area, most notably for ITER at present, where the wins by UK companies presently total nearly 400M. A strong training programme from apprentices, through undergraduate to postgraduate training, is maintained by CCFE. These activities on outreach, industry and training are discussed in section 11.

This is a relatively brief report and contains selected highlights of CCFEs work a list of publications for 2014/15 is given in Annex 1, and these can be referred to for in-depth details of work performed by CCFE. Finally, Annex 2 contains a glossary of technical terms.


Living on the edge

Tokamak physicsHighlight:

Getting peak performance in tokamaks is a trade-off between improving the plasmas confinement without sparking off energy-reducing eruptions at its edge. The two unfortunately go hand-in-hand, but researchers at MAST are looking at ways to increase temperature and pressure at the core of the plasma while delaying the formation of edge instabilities Edge Localised Modes, or ELMs. Using the Beam Emission Spectroscopy diagnostic pictured below, further understanding of ELMs has been achieved. They also found that heating the core meant that a higher pedestal (a high-pressure shield around the plasmas edge which is associated with ELMs) could be achieved before the first instabilities appeared, leading to better overall confinement.

It is the latest discovery by CCFE physicists in the quest to remove or mitigate ELMs; one of the foremost plasma physics problems standing in the way of tokamak development.


4. Tokamak physics

The tokamak physics area covers MAST and JET physics studies and the accompanying theory and modelling programme. The experimental and theory programmes are strongly interwoven. The tokamak physics programme is arranged into four broad areas of work:-

1 Predictable integrated plasma scenarios for ITER, CTF and DEMO. In particular developing high performance operational scenarios on JET with the new ITER-like metal wall and on MAST, preparing scenarios for MAST Upgrade, and developing integrated computational tools to model scenarios.

2 An effective edge pedestal. In the higher performance H-mode the improvement in confinement occurs in the edge region of the plasma, effectively putting the plasma on a pedestal. This area covers formation of a suitable edge pedestal based on physics understanding, and includes development of an understanding of transition of the lower performance L-mode to the H-mode. A key issue is techniques to avoid or mitigating Edge Localised Modes (ELMs), which cause periodic bursts of energy to be released that can give rise to significant heat loads in larger tokamaks.

3 High performance core plasmas with tolerable instabilities. Specifically investigating turbulent energy and particle transport, and controlling it, as a path to higher performance, avoiding or mitigate large scale instabilities and developing understanding of fast-particle driven instabilities and their effects so as to minimise their impact in future devices.

4 Predictive capability to design credible exhaust systems for ITER, CTF and DEMO. In particular to improve understanding of the processes controlling power loads at the plasma-facing components and how they can be reduced in DEMO-class devices the present ITER divertor solution would give rise to intolerable heat loads in DEMO.

We now discuss a few 2014/15 highlights from the tokamak science programme.

Understanding the pedestal on MASTThe pedestal is a region in the edge of a tokamak plasma where the pressure increases rapidly over a very short distance. The pressure at the centre of the tokamak effectively sits on top of this sharp increase, prompting the name pedestal; the larger we can make the pedestal, the hotter we can make the core. Such an improvement does not come for free however, and the presence of the pedestal also gives rise to explosive eruptions of the plasma, similar in nature to solar flares on the sun, called Edge Localised Modes (or ELMs for short). These ELMs carry away some of the energy that has built up in the pedestal out of the plasma, thereby limiting its growth and requiring it to start again. A repeating cycle then begins consisting of pedestal growth followed by an ELM eruption. As a result of this ELM cycle the plasma tends to self-regulate its own core pressure.

During MASTs final experimental campaign before the upgrade, a method for partially bypassing this self-regulation was developed. By delaying the initial growth of the pedestal whilst simultaneously heating the core of the plasma the relationship between the core and the pedestal pressures was temporarily de-coupled. By first increasing the core pressure by 20%, the pedestal pressure was doubled before the first ELM could occur. This seemingly paradoxical effect has two causes: a modification to the plasma profiles as the core pressure is increased and an accumulation of impurities in the pedestal during its growth. The modified profiles delay the onset of the first ELM in the ELM cycle, allowing the pedestal to grow more. As it grows, impurities accumulate, which affects the ELM cycle. Thus, working together, these two effects allow us to generate much hotter cores before the ELM cycle starts.


4. Tokamak physics continued...

The presence of the ELM cycle is ubiquitous to tokamak operation. Despite this the detailed physics underlying the ELM cycle behaviour is still unclear. In MAST light has been shed on this problem through detailed measurements of high frequency fluctuations in the plasma density in the pedestal region throughout the ELM cycle. Measurements of the plasma density during ELM eruptions show a typical blob of density leaving the plasma over tens of microseconds. An example is shown in the left hand figure above. Even more interestingly, prior to the ELM eruption, a fluctuation of the plasma density is seen to occur, shown in the right hand figure above. This fluctuation rotates in the opposition direction to the ELM filaments (see below) and appears to stop rotating just before the ELM eruption occurs. This suggests that the eruption of an ELM is triggered by a change in the rotation of the plasma. If we can understand the ELM trigger mechanism then we may be able to understand how to delay further the onset of the ELM eruption, or even prevent them altogether; a valuable step towards a working fusion reactor.

Integrated ITER-relevant plasmas in JETTokamak plasmas are very delicately balanced. On the one hand the core of the plasma must be hot and dense enough to ensure that fusion occurs; on the other hand, the edge must be cool enough to prevent any significant erosion and damage to materials surrounding the plasma from occurring. In ITER it will be vitally important that integration of these properties is well established if it is to achieve its goal of producing output power ten times greater than its input. In preparation for ITER, JET has been fitted with a new wall consisting of beryllium and tungsten in a configuration mimicking that which will be used when ITER comes online. As a result JET can be used to test and develop operational scenarios in preparation for the challenging environment that will be encountered on ITER.

On ITER it will be imperative that heat fluxes to material components do not exceed 2000 W/cm2 (around 10 to 100 times higher than the heat flux experienced by a space shuttle tile on re-entry). If this limit is exceeded then significant damage to the machine may occur. To ensure that this is not the case gas will be pumped into the tokamak chamber which will be ionised and radiate light as it interacts with the plasma. Through this radiation a large portion of the heat at the periphery of the plasma will be removed and the tokamak materials will survive, relatively unscathed. This then prompts the question, what is the best gas to use, and how does the gas affect the tokamak performance?

On JET nitrogen is the workhorse radiator gas, however its impact on plasma properties is not straightforward. Counterintuitively, the puffing of nitrogen into JET can actually increase the plasma pressure in the pedestal region, despite more power being radiated away. Using nitrogen, JET can now be operated over a wide range of experimental scenarios whilst maintaining the balance between good core performance and tolerable heat loading to materials. Critically, with its ITER-like wall, JET is laying the framework for the operation of ITER.

Understanding the plasma edge through visible imaging on MASTThe periphery of a tokamak plasma is characterized by two regions: the scrape-off layer and the private flux region. The scrape-off layer (SOL) connects the core of the tokamak to material surfaces and so determines how the plasma interacts with materials. The importance of the scrape-off layer cannot be overstated, especially with the issue of excess heat loads on materials being of the highest priority for the operation of ITER. The behaviour of the scrape-off layer is very complex and has warranted extensive research in the past few decades.

The private flux region (PFR) on the other hand, does not connect directly to the tokamak core and is situated between the legs of the divertor (see figure below). Since it does not connect the core to a material surface, the PFR has received muted attention compared to the SOL, with many considering its role in assisting power exhaust in the tokamak to be small. Recent results from MAST are now challenging this perception.

Figure 4.1: Left: A sequence of edge plasma density contours, showing a plasma blob erupting during the ELM. Right: A sequence of edge plasma density contours, just prior to an ELM, indicating a rotating structure.


When the plasma interacts with gas in the vacuum vessel it emits light. Since the majority of this interaction occurs in the SOL and PFR regions, imaging this light can be a very good way of viewing the plasma. In the figure below a camera has been used to image fluctuations in the light emitted from the SOL and PFR during a plasma in MASTs final experimental campaign. The camera records images of the plasma with a frame-rate of 120,000 frames per second. At these speeds it is possible to distinguish individual fluctuations and even analyse their evolution. Rather unexpectedly the PFR contains the most violent fluctuations found in these videos, showing that its nature is significantly more complex than previously thought.

Importantly there appears to be no connection between the appearance of fluctuations in the SOL and the appearance of fluctuations in the PFR, showing that the PFR filaments are truly independent of the rest of the plasma. In fact the filaments are so independent that their behaviour did not change even when the temperature and density of the core plasma were varied. It is hoped that by understanding the nature of these filaments we may be able to exploit them to help reduce the heat and particle loading to material surfaces in ITER and future tokamak fusion reactors. This and other related work will be a focus of the MAST Upgrade programme and accompanying theory, in collaboration with teams from around Europe.

Melting of tungsten by ELM heat loads in the JET divertorThe initial choice of material to use as the main plasma-material interface (called the divertor) in ITER is determined by the need to handle exceedingly high particle and heat loads. Initially, this material was going to be a carbon fibre composite (CFC) to benefit from the many years of experience gained from using CFCs on tokamaks around the world. After this initial phase the CFC on ITER would have been replaced by tungsten prior to full nuclear operation. The success of the JET tungsten divertor in recent years prompted a re-evaluation of this choice and it was proposed to start ITER with a full tungsten divertor resulting in a significant cost and time saving. Tungsten is a metal and, unlike CFCs, may melt under extreme heating events such as ELMs. If ITER is to run smoothly it is very important that we understand how such melting may affect the machine. This has prompted a major experimental campaign in JET to investigate the melting of tungsten due to the impact of ELMs.


Figure 4.2: Fluctuations in the Private Flux Region (PFR) and Scrape-Off Layer (SOL). The broken red lines show the plasma edge and the faint white lines the machine structure.SOL

PFR


A specially designed tile was placed in the divertor in JET. The tile had a raised leading edge which exposed it to significantly harsher heat fluxes than the surrounding tiles, promoting melting localised to that tile. The tile was exposed to seven consecutive plasma pulses aimed at mimicking the conditions likely to occur on the divertor tiles of ITER during operation. The tile temperature reached over 4000oC during ELMs which was enough to cause the tungsten (which has a melting temperature of 3422oC) to melt. The figure above shows the melting of the tile after seven plasma pulses. Interestingly most of the molten material moved along the tile, as seen in the figure, rather than entering into the plasma. A few droplets of tungsten were seen in the JET plasma during these experiments, however they were generally small and had a minimal impact on the plasma itself. This is encouraging because tungsten is a good radiator of light and if too much tungsten accumulates in the plasma it can cause a dramatic reduction in the energy and therefore temperature of the plasma.

A particularly important result from these tungsten melting experiments is the comparison between predictions made using computer simulation and the experimental results. Since the same tools are being used in the design process for ITER the value of a good experimental benchmark cannot be overstated. These results from JET give an optimistic outlook for the operation of ITER with a tungsten divertor, but there are aspects that are not understood so further experiments are planned for JET to increase our understanding of tungsten melting and extend our knowledge base for ITER.

Improved energy confinement in JET with an ITER-like wallTo achieve fusion a tokamak must retain the energy within the plasma for a sufficient amount of time that a significant number of fusion reactions can occur. This process is often referred to as the confinement of energy and is an important factor in preparing scenarios for ITER to achieve its goals of fusion energy gain (more output energy than input energy). The recent installation of an ITER-like wall (ILW) in JET has made it possible to investigate for the first time how energy confinement behaves when the plasma is surrounded by the materials (beryllium for the main chamber and tungsten or the divertor) that will be used in ITER. In order to do this, experiments were conducted where the plasma heating was systematically increased between individual plasma pulses and the confinement of energy measured in each case.


Figure 4.4: Left - Energy confinement of the plasma versus the power input to the plasma. At high power the energy confinement with the ILW is at least as good as that with the former carbon wall, and significantly better than the predictive scaling (IPB98(y,2)) used in the ITER design.


In comparison to the carbon wall, the introduction of the ILW has generally resulted in a lower level of confinement in plasmas with moderate heating power. This is thought to be the result of two factors: A change in composition of the plasma due to a change in wall materials and the use of operational techniques to prevent damage to material components. As the heating power to the plasma is increased however, the story begins to change. For highly-shaped JET plasmas, the JET C-wall experiments show a rapid drop in confinement as the heating power increases. This is in line with predictions based on many years of operating experience. The equivalent JET ILW case does also show a drop in confinement, however the drop is significantly weaker than the C-wall case and, importantly, significantly weaker than predictions suggest. By the time the heating power is increased to maximum, the confinement in the ILW case is similar to that of the C-wall (see Fig 4.4 above). Despite the change in the material used for the wall, careful analysis has shown that this change in the behaviour of the energy confinement is not a result of radiation in the core of the plasma. Instead it is strongly linked to the increase in the energy content of both the core and pedestal regions of the plasma. In plasmas with low shaping a weak drop in confinement is seen with both the C-wall and ILW. In this case the low shaping allowed the plasma to be kept further from the wall at the top of the vacuum vessel, which is thought to reduce the effect of the change in the wall materials on plasma confinement.

These results indicate that wall materials and plasma composition are important factors to be taken into account in the understanding of plasma energy confinement and the prediction of fusion performance in future devices. However, the observation of a weak drop in confinement as the heating power is increased is encouraging for the development of fusion reactors capable of producing high energy gain.



Birth of a tokamak

MAST UpgradeHighlight:

Its not often that you get to see a tokamak being built and the sight of MAST Upgrade taking shape was one of the most fascinating aspects of the year at Culham.The intricately-engineered magnetic coils emerged on cassette assemblies (pictured) just yards from the MAST-U vacuum vessel being prepared to receive them. By the end of the reporting period, the project team were nearing the milestone of joining the first two main parts of the new machine together.

The other photograph shows one highlight from the work the installation of poloidal field coils following a programme of surveying and refurbishment. The coils, which will help shape the plasma inside MAST-U, were carefully lowered backonto strengthened supports within 0.5mm of their optimum position, minimising any stray magnetic fields whenoperations commence.


5. MAST Upgrade

The construction of the new MAST Upgrade (MAST-U) machine has made significant progress during the 2014/15 reporting period. Notable highlights include:

The near completion of the Outer Cylinder of the vacuum vessel and Lower Divertor Nose Cassette with the remaining work on target for them to be connected in June 2015 (see Fig 5.1 below). By April 2015, the main machine assembly was approximately 40% complete;

The assembly of the Lower End Plate of the vacuum vessel is well advanced; Installation and cabling of the majority of the Pulsed Power Supplies in the new MAST-U Power Supplies area, with

local load commissioning of the new Toroidal Field Power Supply on target for commencement at the beginning of May; A successful trial fit of the first of the new Neutral Beam components; The construction of a new Control Room, complete with large Server Room, viewing gallery and in-situ break-out

facility.

Unfortunately a combination of later-than-planned availability of components and an increase in the project cost has required the project schedule to be extended by about one year; the integrated commissioning of the machine is now planned to be in December 2016, leading to initial plasma operation by mid-2017.

Figure 5.1: Cross section of the MAST-U Load Assembly, highlighting sections driving the critical path of the build.

Centre Rod

Lower EndPlate

Vacuum VesselOuter Cylinder

Lower DivertorNose Cassette


Figure 5.1: Cross section of the MAST-U Load Assembly, highlighting sections driving the critical path of the build.

5. MAST Upgrade continued...

The Load Assembly is the core of the MAST-U machine. It comprises the airside coil set, primary vacuum boundary and all load-bearing structures within it such as coils, armour and internal support structures (see Fig 5.1). The Load Assembly has been broken down into five distinct modules for the construction phase and progress for each module is summarised below.

The first Load Assembly module to be installed into the operational area is the Lower End Plate (LEP). This module is expected be completed by December 2015. Key achievements during the past 12 months include: the construction of all 36 coil supports (a total of 6000 parts to assemble); the setting of the three divertor coils to sub-millimetre accuracy; the installation of support structures ready to receive the slats to which the graphite divertor tiles will be attached. The upper end plate is not required for installation onto the vacuum vessel until April 2016 but is already well advanced with numerous coils and key substructures installed.

The outer cylinder is nearing completion and will be ready for the lower cassette to be installed into it by June 2015. The 35 new ports have been welded in position and surveyed. Four of the original Poloidal Field (PF) coils and the support structures for 12 of the original ELM3 coils that are being reused on MAST-U have been installed. The installation of 90 magnetic diagnostics is nearing completion.

The lower Divertor Cassette is virtually complete. The three new divertor coils have been installed on their support brackets to sub-millimetre accuracy. Only the lower of the pair of vertical stabilisation coils (P6) and some associated diagnostic coils remain to be installed before the Divertor Cassette is ready to transfer into the Outer Cylinder module. Work on the upper cassette, which is needed in early 2016, has started.

Work on the Centre Tube module has been a lower priority as this is one of the last modules to be installed and there is significant float relative to the critical path. However, the majority of the diagnostics installed directly to the tube are now in place and all of the Centre Tube armour has been delivered to site.

The 24 wedges that made up the Centre Rod have now been accurately set and overwrapped ready for final impregnation, expected in May 2015. This assembly uses cyanate ester resin rather than the usual epoxy resin to give the required strength at the peak operating temperature. Since the impregnation is the final step in a long manufacturing process, trials using this less familiar resin have been done by the supplier to ensure every step has been validated and reviewed by CCFE before the impregnation is done.

The airside coil set (i.e. those outside the vacuum boundary) comprises the central solenoid and three other coils for plasma shaping and control. As with the Centre Rod, these coils are impregnated with cyanate ester resin and an extensive R&D programme was undertaken with the supplier before the impregnation of the coils. Three coils have now been successfully manufactured.

3ELM = Edge Localised Mode (a plasma instability which these coils are required to mitigate / control; a critical area for magnetic fusion which MAST-U, like MAST, will continue to lead the world in).



Figure 5.2: Progress made on the Lower Divertor Nose Cassette (top left), Lower End Plate (bottom left) andCentre Rod (right).

A key part of the project is improving the reliability and pulse length capability of the Neutral Beam heating systems. To achieve this, two of the critical beamline components are being replaced; namely the Bend Magnets and Residual Ion Dumps.

The majority of the hardware required to upgrade the beamlines is now either on site or about to shipped from the suppliers. As a result, the project is on track to start the mechanical rebuild of both NB systems in early 2016, in parallel with the overhaul of a significant fraction of the Control and Instrumentation infrastructure to operate them.

There are a significant number of diagnostics embedded directly into the Load Assembly modules. With significant constraints on the available space, this has proved more challenging than first thought but the work is progressing well.

In parallel, good progress is being made on the new Thomson Scattering system, a key diagnostic to measure electron temperatures and densities in both the main chamber and new, Super-X divertor, including procurement of the Brewster window (made of high purity glass to produce high laser damage threshold and angled to minimise reflections) and the six large lenses for the collection cell (see Figure 5.3).


Having completed a large proportion of the shielding enhancements to the main MAST-U operational area (known as the blockhouse) previously, the focus for much of the period has been on completing the civil works for the other key operational areas for the new machine. These are now mostly complete, allowing many of the key services, e.g. electrical distribution, network, air conditioning, to be installed.

In parallel, a new Control Room has also been constructed to provide a significantly improved working environment for both home and visiting engineers and scientists (Fig 5.4). It is over four times the size of the previous MAST Control Room, provides a viewing gallery for visitors and, on the mezzanine floor, houses a Server Room with sufficient capacity to cover all future upgrades to the MAST-U machine.


Figure 5.4: Control Room construction phases; exterior (above), interior (below).

Figure 5.3: Progress made on new Thomson Scattering system. Brewster window (bottom left); collection lenses (bottom right).


Gearing up for D-T operation

JET operationsHighlight:

Upgrading JETs detritiation facilities on behalf of our European partners is an important aspect of the preparations to run acampaign of deuterium-tritium experiments in future years. Tritium fuel has to be recovered safely and efficiently from JET as it is both radioactive and scarce. The new facilitiesinclude a new Water Detritiation System to extract tritium from water at Culham and recycle it by way of electrolysis, purification and distillation. This will close the tritium fuel cycle at JET, removing the need for off-site waste disposal.During the year, most of the design work on the facility was completed, and the building itself was constructed readyfor fitting out and commissioning.


6. JET operations

Since 1 January 2014 operation of the JET facilities has been carried out by CCFE under a new bilateral contract between the United Kingdom Atomic Energy Authority and the European Commission. The new JET Operating Contract (NJOC) has a duration of five years until 31 December 2018, and its purpose is to operate and make the JET facilities available to meet the needs of the JET part of the EUROfusion work programme. CCFE also participates in the JET scientific programme as a EUROfusion member, but this section relates to CCFEs role as JET Operator under NJOC.

JET was in the final stages of handover from restart to experimental campaigns at the start of 2014. However, in the first days of January a further technical failure occurred, which necessitated starting an additional immediate intervention. In fact, the impact of this failure was limited by the fact that only one of the two neutral beam heating systems was operational anyway.

An overview of the actual interventions, campaigns and planned shutdown is shown in the following table:

The main CCFE activities are highlighted in this report under the following headings:

Intervention and restart; Campaign highlights; Shutdown activities; Preparation for Deuterium-Tritium operation of JET (DTE2 project); Radioactive waste management.

Intervention and restartDuring the last two days of restart 1 (6-7 January 2014), copper was detected in the plasma, and it was clear that beam operation was causing melting of some copper beam-line components. In addition, a rotary valve on one of JETs neutral beam lines showed abnormal behaviour. A decision was made to suspend operation and to start with the preparation of an intervention to replace the rotary valve and to complete the refurbishment of the other neutral beam systems central support column (CSC) work that was already underway following the earlier problem experienced in October 2013 (reported last year). There were some issues of leak on the CSC during the intervention and in the end pump down was achieved on 17 April.

As reported previously, the planning for the previous major JET restart in 2013 had been optimised and it was completed in less than half the usual time. The schedule for the restart commissioning phase following the above intervention was similarly ambitious, especially since both neutral beam systems had been extensively refurbished. The main aim for this restart phase was to recover plasma operation with maximum neutral beam injection (NBI) power, prior to campaign C33.

Phase Start End

Intervention 13 January 2014 18 April 2014

Restart 18 April 2014 20 June 2014

Campaign C32A 8 January 2014 10 January 2014

Campaign C33 19 June 2014* 5 September 2014

Campaign C34 8 September 2014 9 October 2014

Shutdown 10 October 2014 Mid-June 2015 (plan)

* Restart and Campaign C33 were interleaved for two days


6. JET operations continued...

At the end of the restart phase, the heating systems had achieved 18-20 MW of NBI power, 4 MW of ion cyclotron resonance heating (ICRH) in high-confinement (H-mode) plasmas, and just below 3 MW of lower-hybrid current drive (LHCD) power. Also, the camera protection systems used to monitor and provide real-time protection for plasma-facing components were brought into operation for the solid tungsten tiles and tungsten-coated carbon-fibre composite (CFC) tiles. The restart phase required a total of 50 sessions to complete the commissioning tasks. The total number of pulses per net session was 12.6 for this restart, which is the highest number for a restart phase since installation of the ITER-like wall in 2009 and equivalent to a successful pulse every 30 minutes.

Campaign highlightsThe experiment periods during 2014 are summarised in the table given in the introduction to this section.

C32A: From 8 January 2014 to 10 January 2014The main aim for C32A was to perform experiments with one Neutral Beam Injection system (Octant 4), while the Octant 8 NBI system was being repaired. However, as reported above, on 7 January, the rotary valve of the neutral beam injection system on Octant 4 developed a problem. Hence, C32A was shortened to five shifts of experiments for the period 8-10 January 2014, using ion cyclotron resonance heating only.

C33: From 19 June 2014 to 5 September 2014C33 was very successful and nearly all of the planned experiments were completed. However, the optimisation of high plasma current operation (Ip 4 MA) and the optimisation of the so-called hybrid scenario were reduced in scope due to some limitations experienced in the NBI power. Plasma pulses with NBI were documented in detail for this campaign, in particular the maximum power achieved during the pulse with duration of at least a few hundred milliseconds. This documentation was analysed for specific scenario experiments as summarised in Figure 6.1 below. In conclusion, the maximum power and average power achieved during DT relevant scenario development were 27.2 MW and 19.2 MW, respectively. The design maximum neutral beam power capability is 34 MW. A decision was subsequently made that all PINI beam sources affected by recent vacuum water leaks would be reconditioned on the neutral beam test bed before the start of the 2015 campaigns (see below).

Figure 6.1: Neutral beam performance statistics showing maximum achieved power and average power delivery in the various plasma scenarios studied in Campaign C33 (AT = Advanced Tokamak scenario).



C34: From 8 September 2014 to 9 October 2014C34 was dedicated to a study of hydrogen plasmas, however, during the first day of operation into plasma the NBI box developed problems. After a review, it was decided not to operate the NB systems in hydrogen during C34. Experiments in hydrogen using ohmic plasmas or ICRH were given a total of 26 shifts. A total of eight maintenance days were scheduled for a full regeneration of the cryo-pumps and reconditioning of the machine, to also allow studies of runaway electron suppression by injection of impurity gas of high atomic number (xenon) using a special valve capable of fast, large-quantity gas injection normally used for plasma disruption mitigation. The study of hydrogen plasmas and runaway electron suppression are key experiments in support of preparation for ITER operation.

Shutdown activitiesThe scope of the in-vessel work during the 2014/15 shutdown was originally limited to the necessary tile exchanges inside the torus vacuum vessel for tritium retention and material erosion/migration studies. The shutdown critical path was planned to go through ex-vessel work, dominated by the re-installation of the ITER-like ICRH antenna, and the relocation of the High Frequency Pellet Injector.

A lesson learned from the last major shutdown was the need to carefully complete in a timely fashion Remote Handling (RH) mock-up trials and operator training. Mock-up trials and operator training commenced on 30 April 2014 and was complete before the end of the Campaign C34. This programme resulted in very good reliability of the RH systems during the 2014/15 shutdown, with productivity matching that forecast based on historical levels of in-shutdown maintenance. In fact, the utilisation of RH operational shifts was maintained at 100% throughout, meaning that any unplanned maintenance or remedial work on the RH systems was carried out entirely within the level of contingency included in the schedule.

Following the completion of Campaign C34, the shutdown began on 13 October, 2014 with the major goals of:

Tritium Retention Study Tile Exchange; Re-location of the High Frequency Pellet Injector for improved pellet delivery (shorter flight line); Re-installation of the ITER-like Antenna; Calibrations of visible, infrared and microwave diagnostics; Removal of the Octant 4 NBI Central Support Column assembly for inspection/remedial work and repair; Removal and re-conditioning of all eight PINIs from Octant 4; Replacement of the 36kV switchgear of one of the three main pulsed-power distribution bus-bars; Inspections and refurbishments of penetrations (used for diagnostic access, services etc.) of the main JET biological

shield for DTE2 campaigns; Installation of a third disruption mitigation (massive gas injection) valve; Conversion of the Neutral Beam SF6 towers, housing the high-voltage connection interface, for use with nitrogen as

a replacement insulating gas.

During the shutdown it was agreed to make a manned intervention into the machine in order to inspect and, where possible, repair a number of failed magnetic pick-up coils. This additional work added ~3 weeks to the shutdown schedule and brought the duration of in-vessel activities up to the level of the ex-vessel work.

An important activity at the beginning of all JET shutdowns is an inspection of the state of in-vessel components. This inspection was carried out in January 2015 and identified additional work required to replace tungsten-coated carbon fibre composite divertor tiles and a number of tie-rods located within tiles that had failed due to a manufacturing defect. This added a further month to the shutdown duration.

By the end of the period covered by this report, the planned remedial work on the Octant 4 neutral beam Central Support Column (CSC) had been completed. In parallel to this work on the CSC, a programme of removal, cleaning and reconditioning of 11 PINIs was undertaken. By the end of the reporting period, three PINIs had been reconditioned and the programme remained near the shutdown critical path.



During the reporting period, progress was very good on the 36 kV switchgear replacement project, and the bus-bar was successfully re-energised. As a result of this upgrade, enough fully serviceable units of the original (and now obsolete) switchgear have been released as spares to ensure the remaining bus-bars (that use similar equipment) can be maintained for the foreseeable future.

Progress in the shutdown during the 2014/15 reporting period across all activities has been very good, with delays in the shutdown end date due only to (necessary) extensions to the shutdown scope and not due to problems with system reliability or the availability of components for installation.

Preparation for deuterium-tritium operation of JET (DTE2 project)The main aim of the DT preparation project is to safely prepare the JET machine, its ancillaries and personnel for operations using tritium in experimental campaigns, with the capability to carry out essential machine maintenance and recovery during a DT campaign, the post-DT clean-up and in-vessel sample removal phases. The operator effort for this project includes:

Design, procurement and installation of material and equipment to carry out DT (deuterium-tritium plasmas) and TT Operations (tritium-only plasma) on the JET facilities;

Remote Handling upgrades; Pre-DT shutdown preparation/implementation and restart commissioning; Delivery of the JET Torus Safety Case for DT and TT Operation; Production of revised Operating Procedures for training and certification of operational teams for JET operation with

tritium, evaluated with deuterium; Training of all personnel in emergency scenarios with tritium; Ensuring effective upgrades to tritium accountancy and waste management processes are in place to support tritium

operation.

Figure 6.2: Site of 36kV pulsed-power distribution bus-bar no. 3 following installation of new switchgear units.



During the period of this report the DT preparation project has made significant progress in the main areas identified. The main achievements for the period January 2014 to March 2015 are:

The Project Management Plan was approved and the project developed in both technical and safety case areas. With the identification of detailed tasks and resources, an integrated Project Plan was prepared (baseline September 2014);

Tritium was ordered in November 2014, and a total of 55g will arrive at JET (Active Gas Handling System); The Provisional Facilities Safety Case for DT Operations was endorsed by the Culham Site Safety Working Party.

Although more work is required to achieve the final Facilities Safety Case over the coming two years, this demonstrates that there are no major problems in ultimately achieving this;

The Key Safety Related Equipment / Integrated Operational Protection Systems fitness for purpose study was reviewed at the end of November 2014;

Design and procurement of the Torus Gas Introduction System hardware were underway at the end of the reporting period additional Gas Introduction Modules (GIMs) and Transfer Lines, as well as prototype GIM valves to be tested on JET in 2015;

Characterisation of neutral beam PINI beam source in the neutral beam test bed using the ground-potential grid-gas feed (with deuterium) via which tritium will eventually be fed to the PINI;

Predictive radiation dose maps have been produced as guidance for planning ex-vessel machine access for post-DT activities and at intermediate stages during the campaign if necessary.

Furthermore, a number of additional DTE2 relevant tasks have been added to the project, which have been identified by a recent diagnostic review, and from the recent extensive Reliability Risk Project aimed at identifying high-priority measures for machine reliability risk-mitigation.

Figure 6.3: Elevation at machine octants no. 1 & 5 showing calculated activation dose-rate map, at three months after deuterium-only clean-up phase following planned DT campaign. This indicates the locations, adjacent to the main horizontal ports, where additional measures such as local shielding will be required for manual ex-vessel activities post-DT.



Radioactive waste managementDuring 2014 the significant activities towards reducing the radioactive waste liability and improving the infrastructure to manage waste have continued. The goal remains to minimise the amount of operational radioactive waste on site whilst obtaining best possible value.

In 2014 a new purpose-designed Suited Facility for Bulk Waste (BuS), attached to the Waste Handling Facility (WHF), was completed and commissioned. Double shift operation with two entries per day started at the end of November, and it is foreseen to extend this to three entries per day. Even with the resulting increased throughput, the current projections show that about 16.5t of waste will remain at the end of 2018, requiring a further nine months to clear.

Resources foreseen for activities in the BuS, and the WHF, which was out of operation during the installation of the BuS ventilation system, were diverted to the Unsuited Facility for Bulk Waste (BUF), resulting in a significantly higher than projected throughput for this facility. It is now expected that the backlog of waste for processing in the BUF will be cleared by the start of 2016, freeing resources to start double shift operation in the WHF at that point. The final result was that 30% more waste, across all facilities, was processed in 2014 than had been predicted.

Work has continued on two new facilities, a Water Detritiation System (WDS) to process (rather than export) tritiated water from the Exhaust Detritiation System, and a Material Detritiation Facility to extract tritium from intermediate level waste, allowing it to be classified as low level waste for cheaper and easier disposal. The WDS is under construction with the building complete; however there have been delays with the cryogenic distillation columns requiring a redesign.

Figure 6.4: Active Gas Handling System building J25, with extension (highlighted) nearing completion to house the new Water Detritiation Facility.


Materials investment already proving its worth

Materials researchHighlight:

Studies of materials samples collected from JET after plasma experiments are an early indicator of how Culhams new Materials Research Facility (MRF) can help fusion researchers.

Equipment purchased for the MRF is already in use ahead of the facilitys opening in early 2016. A Thermal Desorption Spectroscopy system has measured the amount of deuterium fuel being retained in JETs plasma-facing surfaces, and dust from the tokamaks divertor area was analysed to assess erosion of materials. The MRF equipment is giving valuable data for European physicists assessing the performance of JETs ITER-like inner wall of beryllium and tungsten.


Our materials research concentrates on (a) how the mechanical properties of structural materials especially low activation steels are degraded by high energy neutrons from fusion, (b) how these and other materials used in fusion become radioactive, and (c) how surfaces in JET are affected by exposure to fusion plasmas. Highlights for 2014/15 include:

Steels in a radiation environment. The suitability of advanced steels for nuclear fusion applications depends on many factors, which determine how their properties change at high temperatures due to exposure to irradiation. One of the key parameters characterising steels is their chemical composition, which determines their fundamental crystal structure. Depending on whether the dominant phase is body centred cubic (bcc) or face centred cubic (fcc) the steels are classified as ferritic (bcc) or austenitic (fcc). One of the fusion steels, Eurofer97, is ferritic. The phase stability of bcc or fcc crystal structures depends on the amount of Cr or Ni added to iron when a steel is made. Chromium alloying favours bcc phase while nickel is an austenite promoting element. In this work, done under an EU funded project led by the European Space Agency, we use first principles density functional theory, treating magnetism of iron, chromium, and nickel, to predict the phases of FeCrNi alloys, deliberately avoiding making any use of experimental data. Calculations spanned more than 500 different alloy structures in the entire alloy composition triangle, for which ab initio calculations were performed. The investigation also involved extensive application of Monte Carlo simulations to model effects of atomic disorder at elevated temperatures. It is the first time that such an extensive fully ab initio analysis of phase stability has been performed for ternary magnetic alloys. The new data enables the development of models for radiation defects in alloys, a necessary step in the assessment of in-service radiation damage effects occurring in steels under fusion or fission operating conditions.

7. Materials research

Figure 7.1: (a) Difference between formation free-energies (eV/atom) of fcc and bcc phases in ternary Fe-Cr-Ni alloys at 600 K. Black solid line separates the Ni-rich region of stability of fcc austenitic alloys from the region of stability of bcc ferritic alloys. (b) Order-disorder temperatures of fcc Fe-Cr-Ni alloys predicted by Monte Carlo simulations.


Materials handbook of activation, transmutation and primary radiation damage properties. Nuclear reaction data is used to predict the neutron irradiation environment and material consequences at various locations within the reactor as a function of time. Experts at the UK Atomic Energy Authority have broad knowledge of how to perform such calculations, with our own, recently updated, inventory software platform called the European Activation System EASY-II, which includes the FISPACT-II inventory simulator. This code suite is used extensively in fusion and non-fusion applications, and is sold commercially to some organisations The nuclear data contained with EASY-II is the most complete and up-to-date available, with uncertainty quantification included, and CCFE is actively involved in its continuing improvement via a set of advanced, extensive, and automated data validation and verification (V&V) systems. This year, the automated infrastructure developed for the V&V reports has been utilised to produce an even more ambitious and complete handbook of activation, transmutation, and radiation damage properties of the elements. This handbook presents inventory simulation results for all naturally occurring elements from hydrogen to bismuth. The handbook now includes predictions of the primary damage events induced in materials by neutrons. The energy distributions of the displaced and recoiling particles produced by nuclear reactions under power plant conditions are crucial input data for modelling the generation of radiation defects. These initial recoil atoms, known as primary knock-on atoms (PKAs), go on to create displacement cascades, which can lead to the life-limiting accumulation of structural damage in a material. Figure 7.2 shows total PKA distributions, which are created by summing the distributions of all of the recoiling species in a given material in iron (Fe), for example, neutrons produce PKAs of Fe, Cr, Mn, He, and H. For comparison, the figure shows that neutron irradiation of SiC produces higher energy PKAs than in Fe. In W, on the other hand, PKAs are comparatively low in energy.

7. Materials research continued...

Figure 7.2: Total PKA distributions (excluding light gas secondary particles) for different elements under DEMO first wall conditions.


Understanding deformation of irradiated metals. Under mechanical stresses, a metal will either accommodate the load in a ductile manner by deforming or in a brittle manner through fracture and cracking. It is vital that the structural metal components in a fusion power plant deform rather than crack, retaining their ductility during operation. Whether a metal is brittle or ductile depends entirely on the motion of defects produced by irradiation, and interaction between the defects. If the defects or dislocations are able to move the metal will deform. If the mobility of defects and dislocations is impeded, it will crack. Defects can be small point defects, formed by only one or several extra or missing atoms, or longer lines and loops, called dislocations. In metals forming the components of a fusion power plant, defects produced by irradiation will spend many years at high temperature, where they are constantly agitated by thermal vibrations of atoms. We have developed new techniques to model how thermal agitation affects the motion of crystal defects, overturning decades of previous theoretical analysis which are then implemented in accurate and efficient simulations. Our current goal is to determine the key processes affecting the ductility of reactor components after many years of high temperature irradiation.

Rotating collector analysis in JET. It is important to know where and how material is transported during plasma operations in JET. Tiles are routinely analysed after each operational period, but these data are integrations over many different plasma conditions. Rotating collectors (RCs) are designed to provide some time resolution by exposing a collector through a narrow slit and rotating the collector a small amount for each plasma pulse data are collected for the first 3000 pulses of the operational period with a resolution of 30-50 pulses (depending on slit width), which is about one day of pulses. A simple application of RCs is to demonstrate the difference in behaviour at the inner and outer divertor between JET operations with carbon plasma-facing components (JET-C) and with metallic surfaces (Be walls and W divertor as will be used for ITER JET-ILW).


Figure 7.3: Left: A line defect (dislocation) in an iron crystal.Right: A small defect cluster and a dislocation with the bulk crystal atoms removed. We have developed novel methods to study the motion of defects stimulated by thermal vibrations of atoms.

Figure 7.4: The positions of the Rotating Collectors in the JET divertor, and the amounts of C and Be deposited in the JET-C and JET-ILW periods. At the inner divertor the amount of Be deposited during the ILW operations is x30 less than the C deposited during JET-C operations, whereas at the outer divertor the reduction is only by a factor of four.


Thermal Desorption Spectroscopy (TDS) of JET. The TDS system that was purchased as part of the National Nuclear Users Facility, in collaboration with the University of Oxford Materials Department, has been used for the analysis of samples from JET. JET samples contain beryllium (Be) and tritium (T), so the system has been installed in a controlled area. The TDS is being used to measure the amount of the plasma fuelling gas (deuterium - D) that is trapped in plasma-facing surfaces, and is a crucial part of experiments to determine the total amount of fuel that is retained during operations; fuel retention will be an important factor in operating future fusion reactors. A total of 15 samples cut from six JET divertor tiles exposed to the ITER-like wall campaign 2010-2012 have been analysed, together with a number of reference samples to calibrate the content. Figure 7.5 shows the count rates for D2 (mass 4) released into a mass spectrometer as a function of time, from two different analysis points. The panel on the left is the result from a sample on Tile 1, whilst the panel on the right is from a position on Tile 6: the exact locations are indicated by the arrows on the divertor cross-section in the central inset. The sample temperature is also shown as a function of the time, so that by comparing the X-intercepts the release temperatures can be assessed the larger D2 peak in the left-hand panel occurs at 345C and the smaller at 625C. The peaks in D2 release from Tile 6 are at similar temperatures, but the relative amounts are very different.

Analysis of dust collected from JET after ILW operations. Understanding dust production in tokamaks remains an important issue for machine operations and safety. Therefore studies of the quantity, composition and structure of dust generated due to erosion of plasma-facing components and disintegrating deposits are part of ongoing tokamak scientific programs. In 2012 dust was collected from the divertor surfaces of the JET vessel following the first operating period of JET-ILW during 2010-2012. Less than 1g of material was collected using a cyclone vacuum cleaner controlled by remote handling. In addition to this large-scale remote collection, samples were taken by hand from smaller areas of a divertor tile. Analysis of the particles collected in this preliminary survey has been carried out in the Materials Research Facility (MRF) at CCFE. The dust samples were collected from a region of high material deposition in the divertor. The particles collected can be divided into three categories: clusters, individual spherical metallic droplets and debris. The secondary electron micrograph in Figure 7.6 shows an example of cluster particles, the most frequently observed particle type. The size of these particles varies in the range 2040 m (as in Fig 7.6) to 44 m. Particles consist of oblong or irregular shape zones with a glassy/amorphous-like appearance (am), fine spherical particles, and crystalline-like agglomerates (cr).


Figure 7.5: Release of D2 (mass 4) as measured by the mass spectrometer and the temperature as functions of heating time for samples cut from a Tile 1 (left-hand panel) and a Tile 6 (right-hand panel). The inset indicates the position of the samples on a divertor cross-section. The amount of D release from the top part of Tile 1 is the largest of any of the cores examined and coincides with the region of greatest Be deposition in the divertor.


Materials Research FacilityAs part of the National Nuclear Users Facility, the Materials Research Facility (MRF) is under construction at the Culham site, with companion facilities in Cumbria (run by NNL and the University of Manchester). This ~9M facility at Culham, funded with 5M from fission funds and the balance from UKAEA, is scheduled to open in late 2015.

The MRF will process active material samples for analysis at the Culham site or at university sites. The capability of the facility to handle active samples is intermediate between universities (very low activity) and Sellafield.

The day one equipment in the MRF, which is already in use for non-activated samples, consists of:

Dual beam Focused Ion Beam which cuts tiny cantilevers for micro-mechanical tests (Fig 7.8), allowing hundreds of experiments to be performed from one sample that is only a few cm long;

A nanoindenter to perform micro-mechanical tests; A scanning electron microscope. This has already been used to make high resolution (~10nm) measurement

of lattice rotations in plastically deformed regions around nano-indents by the technique of Transmission Kikuchi Diffraction (Fig 7.9) the first ever such use;

Thermal Desorption Spectroscopy used to analyse JET tiles (see Fig 7.5 and associated text).

Within the MRF building there is space for further expansion beyond the equipment already procured. Present users of the MRF equipment include the universities of Bristol, Manchester and Oxford, two companies on Culham site, as well as CCFE itself.


Figure 7.6: Secondary Electron Micrograph of a large cluster particle: (am)-amorphous-like zone, (cr)-crystalline-like zone.

Figure 7.7: Architects drawing of the MRF facility (left) and construction of the MRF (right).

Figure 7.8: Cantilevers of a few microns, cut using the Focused Ion Beam.

Figure 7.9: First ever use of Transmission Kikuchi Diffraction for high resolution (~10nm) measurement of lattice rotations in plastically deformed region around nano-indents. Lattice rotation can be converted to strain values. With Oxford Instruments and the University of Oxford.


Bringing 3D printing tofusion research

Technology studiesHighlight:

3D printing is revolutionising the manufacturing of products around the world - from medical implants to jet engine parts. Now the technique, also known as additive manufacturing, is coming to fusion.During the 2014 JET shutdown, three assemblies were remotely installed into the JET vacuum vessel that were manufactured with AM (additive manufactured) parts. The 2B-2D Top Tile Clamps consist of two load-bearing assemblies which attach to the tops of the private limiters that sit either side of the ITER-Like Antenna. The clamps had several parts made from Inconel 718 using AM, as this means of construction was found to be competitive with alternative methods. CCFE performed a qualification program on this material which included extensive mechanical and fatigue testing along with Electron Back-Scatter Diffraction using the Scanning Electron Microscope in Culhams Material Research Facility. The third component is the ITER first mirror test assembly, intended to simulate the ITER first wall diagnostic apertures and consisting of two truncated trapezoidal cones behind which small mirror samples are located. The rather complex surface geometry and thin walls of this plasma-facing assembly meant that AM was the natural manufacturing choice.


Technology studies focus on the design of the prototype power plant, DEMO this work is supported under the Horizon 2020 EUROfusion grant. In addition in the technology area various additive material manufacturing techniques are being explored, partly funded by EU Framework 7 grants.

EUROfusion DEMO StudiesWithin the EUROfusion DEMO studies CCFE secured four of the ten Project Leader roles in Remote Maintenance, Integration, Safety and Environment and Socio-economics with a further two sub-project leaders in Engineering Design Data Integration and First Wall Design. From a total of ten projects, CCFE has involvement in nine, magnets being the only exception. This strong presence reflects CCFEs ability to offer an integrated engineering package from concept design to implementation. A key area of for future development is the Integration project, which will allow CCFE to utilise its full range of expertise.

An example of CCFEs work is on the design of high heat flux components in the DEMO divertor (the region where the majority of the power from the plasma is exhausted). The divertor target is one of the most challenging areas in the design of DEMO, having to sustain tens of MW/m2 surface heat flux as well as aggressive surface erosion and intense 14MeV neutron bombardment. In 2014 CCFE has made further significant strides towards demonstrating a viable DEMO divertor target based on water cooling. The focus is on the CCFE-originated Thermal Break target concept, using a combination of design optimisation, mock-up manufacture and experimental testing.

The divertor design is based on a copper pipe carrying coolant encased in a tungsten block to provide temperature and erosion resistance. Between the pipe and the block is an interlayer and the properties of this interlayer have been the focus of the design optimisation work. Instead of grouping the interlayer properties as a single parameter, the optimisation procedure allows the interlayer elastic modulus (E), coefficient of thermal expansion and thermal conductivity (k) to be independent free variables and separately parameterised.

By treating the three interlayer variables independently, the optimiser finds a large volume of the design space where design reserve factors above 2 are found, i.e. where the structure will not fail (Fig 8.1). The design point labelled in Fig 8.1 passes all the ITER SDC-IC rules and satisfies the supposed 1300C tungsten temperature limit, up to a maximum surface heat flux of 16.5MW/m2. Also plotted is a point representing a FeltMetal interlayer (FeltMetal is metallic cotton wool). Although not quite optimal, FeltMetal is close to the desired region of the response surface and in this design space would yield a minimum reserve factor of ~1.25 (at 10MW/m2). Hence, we conclude that FeltMetal is a viable Thermal/Structural Break material.

The value of this approach is that it allows a rapid assessment and optimisation of designs and material choices without time-consuming experimental tests. It can be applied in many other contexts and any properties can be used to produce a contour surface. As a result of CCFEs work, this methodology has been adopted within the EU DEMO area for future design work and CCFE is coordinating the divertor design effort.

8. Technology studies

Figure 8.1: Response surface plot showing contours of minimum reserve factor as a function of interlayer E and k.


Additive manufacturingThe AMAZE (Additive Manufacturing Aiming Towards Zero Waste & Efficient Production of High-Tech Metal Products) programme is a FP7 project led by the European Space Agency looking at advanced applications of additive manufacturing (3D printing) using metals. CCFE is leading the Fusion High Heat Flux (HHF) working group in collaboration with Birmingham, Swansea and Cranfield Universities (UK), IREPA Laser and Airbus (France), Torino University (Italy) and University of Erlangen-Nuremberg (Germany). The overall aim of the HHF group is the realisation of an advanced divertor target element produced by additive manufacturing techniques; this divertor target element is targeted to operation with high temperature (>600C) coolants to enhance future fusion reactor efficiencies. One of the divertor target element concepts uses tungsten as the armour with a functionally-graded join to a refractory metal structure with millimetre diameter cooling pipes (milli pipes) running throughout.

The use of additive manufacturing is enabling engineers to realise new complex component designs that were previously unattainable due to manufacturing limitations. Adoption of these new complex concepts is opening the opportunity for enhanced performance and the HHF group designs are utilising millimetre pipes through the structure to significantly enhance the cooling capacities of the target element and enabling enhance operational performance.

Within the HHF group the manufacture of multiple complex prototype parts in a range of high temperature materials including vanadium, molybdenum and tungsten has been realised. A half size demonstration part was recently manufactured to prove the capacity to manufacture the complex designs. Figures 8.2 a) to c) shows the manufactured prototype throughout the anticipated manufacturing lifecycle from design, manufactured component to 3D non-destructive analysis of the internal structure (by X-ray tomography). The design and non-destructive analysis for this prototype were performed by CCFE with manufacture at Birmingham University. The result highlights the current success of the group with more advanced designs and divertor target elements manufacturing underway.

The final stages will involve full simulation of the anticipated operational conditions. This simulation will be realised through an experimental campaign focused around testing on a dedicated high heat flux rig being designed and constructed at CCFE.

Addressing fusion materials issues requires that new materials need to be developed specifically to tackle the combined neutron irradiation, heat flux, particle erosion and reduced activation conditions in a fusion reactor. They must also be designed for maximum levels of recyclability and synergy with emerging disruptive manufacturing technologies. A wide range of state-of-the-art analysis, experimental and processing skills and equipment needs to be brought together to form a targeted and unique skills-set for the fast track of design, testing and evaluation of new materials.

8. Technology studies continued...

Figure 8.2: a) CAD model of the prototype component showing conceptual cooling channels; b) photograph of the additively manufactured component; c) X-Ray tomography scan of the manufactured component showing the internal structure of the as-manufactured component.


CCFE has therefore started to place itself at the vanguard of materials technology and developed a strategy that will be taken forward through grand challenges in designing and validating materials for higher strength and fracture under extreme conditions, understanding how manufacturing influences properties, maximising design properties by understanding variability, using strategic elements efficiently and for recyclability.

The divertor and first wall of a fusion reactor will probably be made from tungsten and CCFE has been investigating new methods of producing tungsten components in a rapid and cost effective way; spark plasma sintering offers a potentially attractive manufacturing route. Spark plasma sintering has the potential advantages of enabling a single step manufacturing process (from tungsten powder into net shaped tiles), rapid manufacturing time (several minutes), direct joining capabilities and highly energy efficient production. To date the industrial manufacture of tungsten tiles at appropriate dimensions had not been addressed. Working with leading UK industry, and utilising commercial tungsten powder and an industrial sized spark plasma sintering unit, we manufactured a series of 80mm diameter 10mm height tungsten discs to assess the capacity to manufacture tungsten tiles on an industrial scale.

Critical to the prospect of using the tungsten tiles directly from the spark plasma sintered process, without further time consuming and costly thermo-mechanical treatments, was the level of residual stress in the samples post sintering. To evaluate the levels of residual stress both locally and globally, a series of residual stress testing is underway including X-ray analysis, laser hole drilling and Focused Ion Beam hole drilling. Working in collaboration with the UK Advanced Forming Research Centre catapult at Strathclyde University, the residual stress in the tungsten was evaluated. Figure 8.3 shows one of the tungsten discs produced by spark plasma sintering undergoing X-ray analysis of residual stress. Initial results reveal compressive stresses of the order 200MPa. This relatively low level of compressing stress is a promising indication that these samples could be used in the as-manufactured state.

Ongoing testing utilising the world leading characterisation equipment in the Materials Research Facility and advanced mechanical testing, in collaboration with UK SMEs, will help elucidate the engineering properties of the tungsten samples and enable engineering designs using these new tiles to be evaluated.

8. Technology studies continued...

Figure 8.3: Residual stress investigations of industrially manufactured 80mm diameter spark plasma sintered tungsten disc performed by X-ray analysis at the AFRC.

Industrially manufactured spark plasma sintered tungsten disk


RACE gets off to aflying start

RACEHighlight:

Culhams new remote handling and robotics centre RACE (Remote Applications in Challenging Environments) celebrated its first significant contract soon after opening in Summer 2014.RACE will provide support to an Assystem-led multinational team that has secured a five-year multi-million Euro contractto design, manufacture, install and commission the remote handling equipment for the ITER fusion devices divertorexhaust system. As well as fusion projects, RACE aims to apply its remote handling expertise from JET to other industries requiring similar technology. RACE will soon move to a purpose-built facility on the Culham site equipped to test systems for an array of sectors from oil and gas to nuclear fission and deep space. The building (shown here under construction) will open in early 2016.


A new centre of excellence for Remote Applications in Challenging Environments at the UK Atomic Energy Authoritys Culham Science Centre is now open to users and has already won its first major contract.

RACE has received investment from BIS (UK Government Department for Business, Innovation and Skills) and the Oxfordshire LEP (Local Enterprise Partnership) to conduct R&D into remote applications using purpose built state-of-the-art facilities equipped with remote handling equipment and expertise to design, implement, train and implement complete solutions. End users and their supply chains will be able to develop technology for operations in hazardous environments that include nuclear, oil & gas, sub-sea, space and construction. A number of research areas have been identified:

Manipulators, tools and sensors: The ability to inspect and maintain [The eyes and hands of the remote worker]; Intelligent mobility: The ability to cover large distances to reach the work site [The feet of the remote worker]; Autonomous systems: Using increasing levels of autonomy and improved user interfaces to reduce operator workload

and increase safety and efficiency; Augmented reality: The fusion of advanced modelling and real-time data capture to enhance the operators situational

awareness for planning and doing; Standards: Designing for remote operation and automation requires new standards to be developed to ensure safe

and cost-efficient outcomes.

RACE will focus on demonstrating technology and solutions using mockups, on testbeds within the RACE work hall. RACE will continue to conduct remote handling operations at JET, which is the only fusion reactor with a fully-operational remote handling system.

Projects are underway in an existing laboratory on the Culham site and the new RACE facility will open in 2016. Access is available to all, based on alignment of interests and availability of resources. RACE will bring together SMEs, multinationals, research laboratories and academia in order to create an eco-system that can convert great ideas into real products and value in the market place. The RACE facility brings together a broad range of expertise from the UK Atomic Energy Authority and its partners the National Nuclear Laboratory, The Welding Institute, the National Physical Laboratory and the Nuclear Advanced Manufacturing Research Centre.

9. RACE

Figure 9.1: Architects image of the new RACE building that will be completed at the Culham site in early 2016.


RACE has long-term income streams from JET, DEMO and ITER. Opportunities to develop collaborations, from NNL to China and the US, will ensure that RACE has a strong international dimension and operates at the forefront of technology development.

The RACE team has recently completed a six-month shutdown at JET on schedule, using the remote handling system to undertake a wide range of maintenance activities, including: replacement of a number of in-vessel tile assemblies; divertor Langmuir probes; collection of dust from the divertor for analysis; and calibration of in-vessel diagnostics using a light sphere for spectroscopy measurements (see also section 6).

In addition a number of prototype assemblies were installed an ITER mirror assembly and new ICRH poloidal limiter top adaptor clamp assemblies. These have been produced using additive manufacturing techniques (3D printing) in order to assess their performance in a fusion environment.

RACEs first significant contract is with an Assystem-led consortium developing remote handling technology for ITER. RACE will support design, development, testing and operation of remote maintenance / robotic equipment for one of ITERs key components the divertor exhaust system that ejects waste from the reactor.

9. RACE continued...

Figure 9.2 : ITER mirror assembly being installed in JET.

Figure 9.3 : CAD drawing of the ITER divertor remote handling system (courtesy of Assystem-UK).

Figure 9.4 : ITER Neutral Beam remote handling system [Courtesy of ITER Organization].

In 2013/14 the RACE team completed the conceptual design of the ITER Neutral Beam Remote Handling System and supported industry in the final proposal for complete supply and installation.

These two contracts show how RACE intends to work by supporting businesses, especially UK firms, to wincommercial contracts.


Successful conclusion to neutral beam design work

ITER systemsHighlight:

CCFE, along with Spanish fusion association Ciemat, has carried out much of the design work to integrate ITERs main plasma heating system (the neutral beam injector) into the rest of the tokamaks assembly. Although the neutral beam technology is

euratom/ccfe fusion association annual report … · euratom/ccfe fusion association annual report...

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